Radiation Survey Process

ABSTRACT

A method for determining a radionuclide concentration of a material is provided. The method comprises placing the material to be analyzed into a vessel, wherein the material comprises a radionuclide, wherein the material has a known volume, and wherein the vessel has a fixed geometry. The method further comprises weighing the material to be analyzed and measuring the moisture content of the material to be analyzed. The method additionally comprises placing a protective structure in the material and placing a detector in the protective structure, wherein the detector is coupled to a single-channel analyzer. The method also comprises counting the emitted radiation having a known energy over an interval of time to produce a count per time, wherein the emitted radiation is emitted from the radionuclide and then dividing the count per time by the weight of the material to produce a count per time per weight.

CROSS-REFERENCE TO RELATED APPLICATIONS

This application claims priority to U.S. Application Ser. No. 62/065,239filed on Oct. 17, 2014, and to U.S. Application Ser. No. 62/072,424filed on Oct. 30, 2014, the disclosures of which are herein incorporatedby reference in their entirety.

BACKGROUND OF THE INVENTION

1. Field of the Invention

This invention relates to the field of radionuclide detection andcharacterization; and more specifically to the quantification ofradionuclides by measurement of single-energy gamma rays in afixed-geometry system.

2. Background of the Invention

A number of industries, including oil and gas production, rare earthmineral processing, and phosphate rock processing, generate wastecontaminated with radioactive materials. These radioactive materials aregenerally referred to as naturally occurring radioactive materials(NORM) or technologically enhanced naturally occurring radioactivematerials (TENORM). NORM and TENORM are not man-made radioactivematerials and are not produced by splitting atoms in a nuclear reactoror by bombarding atoms with subatomic particles in accelerators. In thecase of TENORM, the radionuclides may be enhanced by man, for example,by concentration of the radionuclides, but like NORM, TENORM is notproduced by man. Some of the radioactive materials in NORM and TENORMmay be subject to environmental regulations and require specialdisposal. NORM and TENORM may include radionuclides such as U-238,U-235, Th-232 as well as their respective progeny radionuclides (e.g.,Ra-226, Po-210, etc.). As U-238, U-235, Th-232 and the unstable isotopesof their progeny decay; alpha, beta, and gamma radiation may be emitted.Radiation may continue to be emitted until a stable isotope in the chainof decaying progeny is reached, for example, Pb-206 is a stable isotopeand the end product of the U-238 decay chain. The emitted radiation fromthe unstable radionuclides may be detected and used to calculate theconcentration and activity of the unstable radionuclides, and moreparticularly, those unstable radionuclides which are subject toenvironmental regulation.

Present techniques for the detection of NORM and TENORM involveextracting samples from a suspect material (e.g., a waste) and thensending each individual sample to a laboratory for analysis, orperforming gamma spectroscopy in the field. Laboratory analysis may becostly in both time and expense. For example, a typical laboratoryanalysis of a sample may take weeks and require expensive and rigoroussample processing. Further, during the time in which a sample is beinganalyzed, the remainder of the suspect material may be moved to andstored in a holding facility until the laboratory has completed thesample analysis. Also, as discussed above, each and every sample may besent to the laboratory for analysis. Thus, any time a new waste materialis produced or a NORM/TENORM survey is needed for any other reason,samples may be extracted and sent to the laboratory for analysis.

Alternatively, field analysis may be done using specialized equipment.For example, the activity of radionuclides may be measured using amulti-channel analyzer that scans a wide range of gamma ray energies,allowing for the identification and quantification of photons withenergies corresponding to the applicable radionuclide progeny. This,multi-channel analysis for gamma radiation (gamma spectroscopy) mayrequire careful sample preparation, and may include drying the sample,weighing a measured amount for analysis, and counting using a gammaspectroscopy system for a time ranging from several minutes to manyhours. This technique is complex and may require an involved samplepreparation procedure. Further, multi-channel analyzers used to capturemultiple emitted gamma rays at once are expensive. Lastly, the operationof multi-channel analyzers is complex and may require trained personnelto operate in order to produce accurate measurements, for example, someradionuclides may emit radiation of an energy which obscures or may beconfused with the emitted radiation of another radionuclide desired foranalysis. Without proper training, it may be difficult to distinguishbetween the two activities of the radionuclides and thusly, aninaccurate measurement may be produced.

Consequently, there is a need for an improved radiation survey processfor a practical cost-effective method of quantifying the amount ofradionuclides in a waste sample in the field.

BRIEF SUMMARY OF SOME OF THE PREFERRED EMBODIMENTS

These and other needs in the art are addressed in an embodimentcomprising a method for determining a radionuclide concentration of amaterial. The method comprises placing the material to be analyzed intoa vessel, wherein the material may be contaminated with one or moreradionuclides, wherein the material has a known volume, and wherein thevessel has a fixed geometry. The method further comprises weighing thematerial to be analyzed and measuring the moisture content of thematerial to be analyzed. The method additionally comprises placing aprotective structure in the material and placing a detector in theprotective structure, wherein the detector is coupled to asingle-channel analyzer. The method also comprises counting the emittedradiation having a known energy over an interval of time to produce acount per time, wherein the emitted radiation is emitted from theradionuclide and then dividing the count per time by the weight of thematerial to produce a count per time per weight.

These and other needs in the art are addressed by an additionalembodiment comprising a system for characterizing the radionuclidecontent of a material. The system comprises a vessel having a fixedgeometry, wherein the vessel is capable of containing a material thatmay be contaminated with one or more radionuclides. The system furthercomprises a protective structure inserted into the vessel and a detectorinserted into the protective structure, wherein the detector is coupledto a single-channel analyzer.

The foregoing has outlined rather broadly the features and technicaladvantages of the present invention in order that the detaileddescription of the invention that follows may be better understood.Additional features and advantages of the invention will be describedhereinafter that form the subject of the claims of the invention. Itshould be appreciated by those skilled in the art that the conceptionand the specific embodiments disclosed may be readily utilized as abasis for modifying or designing other embodiments for carrying out thesame purposes of the present invention. It should also be realized bythose skilled in the art that such equivalent embodiments do not departfrom the spirit and scope of the invention as set forth in the appendedclaims.

BRIEF DESCRIPTION OF THE DRAWINGS

These drawings illustrate certain aspects of some of the embodiments ofthe present method, and should not be used to limit or define themethod.

FIG. 1 illustrates an example embodiment of a method of quantifying theconcentration of a radionuclide contained within a target material; and

FIG. 2 illustrates another example embodiment of a method of quantifyingthe concentration of a radionuclide contained within a target material.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

This invention relates to the field of radionuclide detection andquantification for appropriate disposal; and more specifically, to thequantification of radionuclides by measurement of single-energy gammarays in a fixed-geometry system. In embodiments, a single-channelanalyzer may be used to detect the presence of NORM and TENORM in amaterial (e.g., a waste material). In embodiments, a single-channelanalyzer may count a gamma ray of a specific energy and in a fixedgeometry such that the concentration and activity of a radionuclide maybe determined. In embodiments, the activity of the radionuclide maydetermine the method of disposition of the material.

Embodiments comprise a single-channel analyzer. The single-channelanalyzer may be any single-channel analyzer sufficient for capturingemitted gamma energy. Examples of a single-channel analyzer may include,but are not limited to scintillation-types such as sodium iodide,lanthanum bromide, cadmium-zirconium-telluride; or solid state typessuch high-purity germanium detectors. In some embodiments, gas-filleddetectors, for example, Geiger-Mueller or ion chamber detectors may notbe sufficient for quantifying the concentration of a radionuclide. Moreparticularly, Geiger-Mueller detectors may not be viable because they donot have an energy-dependent response and as such, they cannotdistinguish between high-energy and low-energy gamma emissions.Proportional counters may not be viable because their response isdependent upon detector voltage, which may change as batteries aredepleted or replaced, thus significantly impacting their consistency inthe field. Ionization chambers may not be viable because their detectionefficiency is too low for the applications disclosed herein. Thesingle-channel analyzer may be used to measure a specific window ofgamma energy. For example, the specific window of measurable gammaenergy may comprise a range encompassing 0.609 MeV. In this example, areading of 0.609 MeV corresponds to a gamma ray emitted by Bi-214, whereBi-214 is a progeny in the decay chain of U-238. This radiation energyhas an intensity of 46.1%, which refers to the percentage ofdisintegrations of the nuclide itself in which a gamma ray of thisenergy is emitted. Thus, a 0.609 MeV is emitted in the decay of Bi-21446.1% of the time. By narrowing the specific window to a encompass anarrow range of emitted radiation energies, the single-channel analyzermay be used to quickly identify the presence of a targeted nuclide,which may in turn be used to determine the concentration and activity ofthe targeted radionuclide. Further, the targeted radionuclide may beused to determine the concentration of other radionuclides in the decaychain in which the targeted radionuclide is a link.

Embodiments comprise a method of quantifying the concentration of aradionuclide in a target material. A target material is any material inwhich it may be desirable to measure the concentration and/or activityof the radionuclides disposed within. The target material may be a wastematerial. The target material may be a material produced through oil andgas production, rare earth mineral processing, phosphate rockprocessing, the like, or any combinations thereof. The target materialmay comprise a solid or may comprise a mixture of a solid and liquid. Inembodiments, the target material may comprise nuclides and/orradionuclides. Target materials comprising radionuclides may emitradiation energy as the radionuclides decay. For example a targetmaterial comprising Ra-226 may emit gamma radiation of 0.186 MeV.

Embodiments comprise calibrating the singe-channel analyzer to detect aradionuclide of interest. For example, if the single-channel analyzer isused to detect Bi-214, the single-channel analyzer window may be set todetect the energy of a 0.609 MeV photon of emitted gamma radiation. Thismay be accomplished by holding the detector portion (e.g., ascintillation crystal) of the single-channel analyzer next to aradioactive source that emits the radiation energy of interest, orapproximately the radiation energy of interest (e.g., the 0.609 MeVgamma ray emitted by Bi-214 as described above). The single-channelanalyzer may then be adjusted to a desirable energy threshold, inembodiments, this adjustment may be performed through hardware,electronic, and/or software adjustment. For example, a thresholdpotentiometer may be adjusted such that the detector energy threshold is20 keV below the radiation energy of interest. Alternatively, thedetector energy threshold may be set to 30 keV below the radiationenergy of interest, or further alternatively, the detector energythreshold may be set to 10 keV below the radiation energy of interest.Adjusting the detector energy threshold of the threshold potentiometeris within the capabilities of one of ordinary skill in the art and maybe adjusted as desired for a specific application. Further calibrationmay be done by adjusting the single-channel analyzer counting window tothe desired amount, for example the counting window may be adjusted to40 keV using the window potentiometer to exclude counting any gamma raysexcept for those in the counting window. After adjustment, the detectorof the single-channel analyzer may be held to the radioactive sourceagain. This time, the count of emitted gamma rays above the thresholdmay be taken. The count rate may be recorded every 5 volts or asdesired, for example every 1 volt, every 2 volts, every 3 volts, every10 volts, and the like. The detector voltage may then be set at thelevel that gives the highest count rate, or alternatively that is in themiddle of a plateau in the count rate. A plateau in the count rateoccurs when the count rate changes very little with increasing voltage.If the count rate differs from the average value when counting a sourceof known strength (e.g., during a daily instrument response check) bymore than 20% the meter may require recalibration or repair.

The single-channel analyzer may be used to measure the emitted radiationof any radionuclide. The radionuclide may be any decay series nuclide.For example, the radionuclide may comprise Bi-214, Ra-226, Pa-234, andthe like. Factors to consider when selecting an energy window mayinclude the intensity of the radiation energy emitted (i.e. thefrequency with which gamma ray photons of a particular energy areemitted), the energy of the emitted radiation, the half-life of theisotope, whether the single-channel analyzer and consequently detectortype chosen can measure the radiation energy, other radionuclides in thesame decay chain, as well as other factors which would be apparent toone of ordinary skill in the art.

Embodiments comprise the measurement of background radiation at the siteof the target material. Background radiation may interfere with theaccuracy of the quantification of the concentration of a radionuclide.Thus, embodiments comprise measurement of the background radiation sothat it may be subtracted from the gross count of radiation of thetarget material. In order to measure the background radiation, thesingle-channel analyzer may be moved away from the target material suchthat the majority or all of the emitted radiation energy from the targetmaterial will not be detected by the single-channel analyzer. At asufficient distance from the target material, the single-channelanalyzer may be used to count the background radiation for a specifictime interval, for example, 10 minutes. The time interval may be chosenas desired and selection of a time interval to perform a backgroundradiation count is within the purview of one of ordinary skill in theart.

In optional embodiments, a radiation safety survey may be performed. Theradiation safety survey may be performed using the single-channelanalyzer or any other type of analyzer, including those that may beineffective for the methods of quantifying the concentration of aradionuclide as disclosed herein, for example, Geiger-Mueller or ionchamber detectors. The radiation safety survey may be done to ascertaina general level of radiation emitted by the target material such thatappropriate safety protocols are utilized in the handling of the targetmaterial. In these optional embodiments, the analyzer may be calibrated,either as described above, or in any other appropriate manner if theanalyzer is not already calibrated. Once the analyzer is calibrated, abackground radiation count may be performed as disclosed above or anyother sufficient manner. Once the background count has been obtained,the background count may be noted so that it may be removed, zeroed out,or otherwise subtracted from the measurement of the radiation count ofthe target material. For the radiation safety survey of the targetmaterial, a survey location near the target material may be selected.Once a survey location is selected, the analyzer may be used to measurethe radiation count (i.e., the radiation count above background) at thesurvey location. This process may be repeated at multiple surveylocations if desired. Appropriate safety measures, including protectiveclothing and isolation of the target material may be required should thetarget material emit an amount of radiation energy above a threshold inwhich one of ordinary skill in the art would deem such measuresappropriate. Such threshold may be provided by environmental regulationsor may otherwise be known to one of ordinary skill in the art. Furthershould the target material emit a level of radiation indistinguishablefrom the background or otherwise at a level that does not exceed athreshold set by environmental or any other type of regulation (i.e.,the target material may be deemed to not comprise NORM/TENORM), then insome circumstances it may be possible to dispose of the target materialwithout further testing.

FIG. 1 illustrates an example embodiment of a method of quantifying theconcentration of a radionuclide disposed within a target material. Withreference to FIG. 1, a vessel 5 is filled with target material 10.Vessel 5 may be any such vessel having a fixed geometry and capable ofcontaining target material 10. Vessel 5 may comprise any shape having afixed geometry. In embodiments, the vessel 5 has any shape suitable toallow the detector 20 to be placed at about the approximate center ofvessel 5 and/or target material 10. In some embodiments, the vessel 5has a cylindrical shape or a hemispherical shape. In an embodiment,vessel 5 has a cylindrical shape. Vessel 5 may be made of any materialsuitable for containment of a target material 10 which may compriseradionuclides. Examples of materials may include plastic, wood, metal,the like, composites thereof, or combinations thereof. Target material10 may be obtained from any material in which a quantification of theconcentration of one or more radionuclides is desired. Target material10 may be a waste material. In embodiments, when vessel 5 is filled withtarget material 10, the amount of void space within vessel may beminimized, such that the vessel 5 is filled with target material 10 toas great an extent as reasonably possible. Further, in embodiments, astandard vessel 5, with a standard volume filled to a standard level andwith a standard counting geometry may assist in maintaining consistencyin the measurements. This may be of particular importance when averagingseveral sample measurements. As such, more accurate and consistentmeasurements may be obtained through the use a standard vessel 5 with astandard volume filled to a standard level and with a standard countinggeometry. Further, should target material 10 comprise a solid mixed witha fluid, for example water, the fluid may be drained to preventinterference with the measurement. The fluid may be drained by anysuitable method. Once the target material 10 is disposed within vessel5, a protective structure 15 may be inserted into the target material 10disposed within vessel 5. In embodiments, protective structure 15 may beinserted as close to the fixed geometric center of the target material10 disposed within vessel 5 as reasonably possible, such that the amountof the target material 10 on any side of the protective structure 15 isapproximately equal to the amount of the target material 10 on theopposing side of the protective structure 15. Additionally, theprotective structure 15 may be placed such that a sufficient amount ofthe target material 10 surrounds all sides of the protective structure15. In some embodiments, three inches or more of the target material 10or more may surround all sides of the protective structure 15. In someembodiments, a range of about two inches to about four inches of targetmaterial 10 may surround all sides of the protective structure 15. Inother embodiments, a range of about three inches to about four inches oftarget material 10 may surround all sides of the protective structure15. In still further other embodiments, a range of about two inches toabout six inches of target material 10 may surround all sides of theprotective structure 15. In alternative embodiments, four inches or moreof the target material 10 may surround all sides of the protectivestructure 15. In further alternative embodiments, five inches or more ofthe target material 10 may surround all sides of the protectivestructure 15. With the benefit if this disclosure, one of ordinary skillin the art will be able to select an amount of the target material 10 tosurround the protective structure 15.

In embodiments, protective structure 15 may be made of any low densitymaterial. In some embodiments, protective structure 15 may provide thedetector 20 an approximate 4π counting geometry. Protective structure 15may protect detector 20 from contact with the target material 10 suchthat contamination may not occur. However protective structure 15, maynot protect detector 20 from contact with any emitted radiation from thetarget material 10. Thus, the materials used for protective structure 15may comprise a density low enough to allow as much emitted radiation topass through protective structure 15 and to detector 20 as reasonablypossible. Examples of protective structure 15 materials may includewood, polyvinyl chloride, aluminum, the like, combinations thereof,and/or any composites thereof. Calibration of detector 20 and acorresponding single-channel analyzer may be performed if suitable asdescribed above. Detector 20 may be inserted into the protectivestructure 15 such that the detector 20 is placed as close to the fixedgeometric center of the target material 10 disposed within vessel 5 asreasonably possible. Detector 20 may be coupled to a single channelanalyzer. Detector 20 may comprise any type of detector 20 suitable fordetecting a radiation energy emitted from a radionuclide. In anembodiment, detector 20 comprises a radiation detector comprisingscintillation crystals. In embodiments, detector 20 comprises cesiumiodide, sodium iodide, lanthanum bromide, cadmium-zirconium-telluride,lithium iodide, high-purity germanium, the like, or any combinationsthereof.

With reference to FIG. 2, single-channel analyzer 25 may be coupled todetector 20. Single-channel analyzer 25 may be used to count the emittedradiation energy from the target material 10. With detector 20 in placewithin protective structure 15, single-channel analyzer 25 may be usedto count a window of emitted radiation energy. For example,single-channel analyzer 25 may measure an energy window set to includean emitted gamma photon of 0.609 MeV, such an emission corresponds to adecay of an atom of Bi-214. Once the window is set, the single-channelanalyzer 25 may then count emitted radiation energy within the windowfor a time interval as desired to provide a count per time. For example,the single-channel analyzer 25 may perform a count for 1 minute, 2minutes, 3 minutes, 4 minutes, 5 minutes, or longer as desired. With thebenefit of this disclosure, one of ordinary skill in the art will beable to select a timer interval for performing a count for a specificapplication. After a count is performed, the counting process may berepeated on subsequent samples as desired. Once the counts are obtained,the background radiation may be subtracted from counts to produce thenet count rate which may be referred to as “counts above background.”Further the weight and moisture content of the target material 10 may bemeasured during the counting process such that the weight and moisturecontent of the target material 10 are reflective of the time in whichthe count was performed. The weight of the target material 10 ismeasured in order to compute a count per time per mass such as thecounts per minute per mass (“CPM/M”). The moisture content may be usedto correct the activity concentration of the measured radionuclide asdescribed below. In embodiments, measuring the weight and moisturecontent of the target material 10 at the time in which the counts aretaken may produce a more accurate measure of CPM/M as compared tomeasuring the weight and moisture content of the target material 10 at alater time in which the target material 10 may have drained or otherwisechanged in weight and/or moisture content, such that the weight and/ormoisture content of target material 10 is no longer reflective of theweight and/or moisture content of target material 10 at the time inwhich the counts were obtained.

Once the single-channel analyzer measurement of CPM/M is produced, theCPM/M may be converted to an activity measurement, which is a measure ofradioactivity per mass. For example, if the mass of the CPM/M is inkilograms, the CPM/kg may be converted to pCi/gm. In embodiments, theconversion of the CPM/M to an activity measurement includes multiplyingthe CPM/M by a conversion factor. In this procedure, it is assumed thatthe radioactive decay series nuclides are in secular equilibrium so thatthey are all present in roughly equal activity concentrations. Secularequilibrium may only occur when the half-life of the unstable parentisotope is much longer than the half-life of the unstable daughterisotope. In such a case, the unstable daughter isotope emits radiation(i.e., decays to its daughter isotope) more quickly than the unstableparent isotope. Therefore, with a sufficient amount of time, theunstable daughter isotope will decay at the same rate it is produced.Thus, if all of the radionuclides in a decay chain are in secularequilibrium, the quantification of one radionuclide and consequently theactivity of the same radionuclide allows for the quantification andactivity of other radionuclides within the decay chain. For example, ina target material 10, such as a sample of NORM, the activity measurementof Bi-214 (or any other nuclide in the U-238 decay series) allows forthe determination of the activity measurement of Ra-226. This may be ofspecial importance in target materials 10 comprising NORM. As discussedabove, NORM comprises non-man-made radionuclides; these primordialradionuclides (e.g., U-238, U-235, and Th-232) may approximately be in astate of secular equilibrium due to their age.

Unlike NORM, TENORM, although not man-made, may not be in secularequilibrium in all cases due to industrial activities which may breakthe secular equilibrium chain, for example, through processing to removeand/or concentrate various unstable isotopes in the decay chain. In suchcases, a conversion factor that accounts for the variables noted aboveplus the age of the sample (i.e. the amount of time since the secularequilibrium was disturbed), may be used to determine the current stateof the radionuclides present relative to the state of the radionuclidesas they would be in a decay chain in secular equilibrium. The conversionfactor thusly allows for correction in the sample variability. Theconversion factor may be obtained by calculation and/or by comparinglaboratory test results to field readings. A typical laboratory test toobtain the conversion factor may be gamma spectroscopy. Gammaspectroscopy may be used to identify the activity of all gamma emittersand thus, whether the gamma emitters are in secular equilibrium. Withthe benefit of this disclosure, one of ordinary skill in the art will bereadily able to determine an appropriate conversion factor for a targetmaterial 10.

The laboratory analysis to obtain the conversion factor may, in someembodiments, need only be obtained once, provided the sample of targetmaterial 10 from which the conversion factor was obtained is the same assubsequent samples of target material 10 obtained from the field. Forexample, any subsequent samples of NORM may or may not be able to use aconversion factor obtained from a NORM sample taken from a differentsite, depending on the characteristics of the material being sampled andtheir similarity to known samples. In this case, the sample of NORM inwhich the conversion factor was determined may be different from thesubsequent NORM samples, and therefore the conversion factor may notprovide a proper conversion. As a further example, samples of TENORMthat have undergone processing in which secular equilibrium is broken,may not use a conversion factor obtained from a sample of TENORM/NORM inwhich the processing did not occur, even if the samples were obtainedfrom the same site. In some cases, secular equilibrium may bereestablished with time to correct these issues, as would be apparent byone of ordinary skill in the art, in which case a prior obtainedconversion factor may again be used. Once the conversion factor isobtained, all field samples for which it is relevant may then bemeasured according to the methods disclosed herein, and may be done sowithout the need to send each subsequent sample to a laboratory fortesting or to use complex and expensive equipment such as multi-channelanalyzers. Further, the above described methods may be repeated asdesired on as many samples as desired in order to obtain an averagevalue of the activity.

After the CPM/M has been multiplied by the conversion factor to producethe activity concentration, correction of the activity concentration maybe desired by applying the appropriate moisture and density correctionfactors as would be done by one of ordinary skill in the art. Theconverted and corrected activity concentration of the radionuclidespresent in the target material 10 may then be used to determine thedisposition of the target material 10.

As discussed above, the appropriate disposition method of targetmaterial 10 may be determined based on the methods described above. Theappropriate method of disposition may largely be determined by theactivity concentration of specific radionuclides and any regulationsregarding those specific radionuclides. For example, if an environmentalor any other type of regulation restricts an activity concentration forRa-226 to less than 5 pCi/gm, a measured average activity concentrationof 4 pCi/gm for Ra-226 allows for disposal of the target material 10without regards to the radioactivity present. Conversely, should themeasured average activity concentration of Ra-226 be 7 pCi/gm, thetarget material 10 may be disposed of as NORM/TENORM waste according tothe controlling regulations of the site in which the target material 10is located. Disposal of NORM/TENORM typically entails disposal at a siteapproved for NORM/TENORM waste.

Although the present invention and its advantages have been described indetail, it should be understood that various changes, substitutions andalterations may be made herein without departing from the spirit andscope of the invention as defined by the appended claims.

It should be understood that the compositions and methods are describedin terms of “comprising,” “containing,” or “including” variouscomponents or steps, the compositions and methods can also “consistessentially of” or “consist of” the various components and steps.Moreover, the indefinite articles “a” or “an,” as used in the claims,are defined herein to mean one or more than one of the element that itintroduces.

For the sake of brevity, only certain ranges are explicitly disclosedherein. However, ranges from any lower limit may be combined with anyupper limit to recite a range not explicitly recited, as well as, rangesfrom any lower limit may be combined with any other lower limit torecite a range not explicitly recited, in the same way, ranges from anyupper limit may be combined with any other upper limit to recite a rangenot explicitly recited. Additionally, whenever a numerical range with alower limit and an upper limit is disclosed, any number and any includedrange falling within the range are specifically disclosed. Inparticular, every range of values (of the form, “from about a to aboutb,” or, equivalently, “from approximately a to b,” or, equivalently,“from approximately a-b”) disclosed herein is to be understood to setforth every number and range encompassed within the broader range ofvalues even if not explicitly recited. Thus, every point or individualvalue may serve as its own lower or upper limit combined with any otherpoint or individual value or any other lower or upper limit, to recite arange not explicitly recited.

Therefore, the present embodiments are well adapted to attain the endsand advantages mentioned as well as those that are inherent therein. Theparticular embodiments disclosed above are illustrative only, and may bemodified and practiced in different but equivalent manners apparent tothose skilled in the art having the benefit of the teachings herein.Although individual embodiments are discussed, the invention covers allcombinations of all those embodiments. Furthermore, no limitations areintended to the details of construction or design herein shown, otherthan as described in the claims below. Also, the terms in the claimshave their plain, ordinary meaning unless otherwise explicitly andclearly defined by the patentee. It is therefore evident that theparticular illustrative embodiments disclosed above may be altered ormodified and all such variations are considered within the scope andspirit of the invention. If there is any conflict in the usages of aword or term in this specification and one or more patent(s) or otherdocuments that may be incorporated herein by reference, the definitionsthat are consistent with this specification should be adopted.

What is claimed is:
 1. A method for determining a radionuclideconcentration of a material comprising: placing the material to beanalyzed into a vessel, wherein the material comprises a radionuclide,wherein the material has a known volume, and wherein the vessel has afixed geometry; weighing the material to be analyzed; measuring themoisture content of the material to be analyzed; placing a protectivestructure in the material; placing a detector in the protectivestructure, wherein the detector is coupled to a single-channel analyzer;counting the emitted radiation having a known energy over an interval oftime to produce a count per time, wherein the emitted radiation isemitted from the radionuclide; and dividing the count per time by theweight of the material to produce a count per time per weight.
 2. Themethod of claim 1, wherein the radioactive material comprises NORMand/or TENORM.
 3. The method of claim 1, wherein the vessel iscylindrical.
 4. The method of claim 1, wherein the protective structurecomprises wood, polyvinyl chloride, aluminum, or any combinationthereof.
 5. The method of claim 1, wherein the single-channel analyzeris a scintillation-type or solid-state type.
 6. The method of claim 1,wherein the detector comprises a radiation detector, wherein theradiation detector comprises cesium iodide, sodium iodide, lanthanumbromide, cadmium-zirconium-telluride, lithium iodide, high-puritygermanium, or any combinations thereof.
 7. The method of claim 1,wherein the radionuclide is Bi-214.
 8. The method of claim 1, whereinthe emitted radiation comprises gamma radiation.
 9. The method of claim1, wherein the count per time per weight is multiplied by a conversionfactor to produce an activity concentration of the radionuclide.
 10. Themethod of claim 9, wherein the conversion factor is obtained by gammaspectroscopy of the material.
 11. The method of claim 1, wherein themethod is repeated to obtain multiple counts per time per weight,wherein the multiple counts per time per weight are averaged to producean average count per time per weight.
 12. A system for characterizingthe radionuclide content of a material comprising: a vessel having afixed geometry, wherein the vessel is capable of containing a materialcomprising a radionuclide; a protective structure inserted into thevessel; and a detector inserted into the protective structure, whereinthe detector is coupled to a single-channel analyzer.
 13. The system ofclaim 12, further comprising the material comprising a radionuclide. 14.The system of claim 13, wherein the radionuclide is Bi-214.
 15. Thesystem of claim 13, wherein the radionuclide is capable of emittinggamma radiation.
 16. The system of claim 13, wherein the materialcomprises NORM and/or TENORM.
 17. The system of claim 12, wherein thevessel is cylindrical.
 18. The system of claim 12, wherein theprotective structure comprises wood, polyvinyl chloride, aluminum, orany combination thereof.
 19. The system of claim 12, wherein thesingle-channel analyzer is a scintillation-type or solid-state type. 20.The system of claim 12, wherein the detector comprises a radiationdetector, wherein the radiation detector comprises cesium iodide, sodiumiodide, lanthanum bromide, cadmium-zirconium-telluride, lithium iodide,high-purity germanium, or any combinations thereof.